The impending energy crisis has been occupying considerable attention of scientists and power engineers in recent years. The dependence on nuclear fuels for the future appears more than likely. ("Power in the Year 2001"--Mechanical Engineering, December 1971.)
Of the well known fissionable isotopes used for nuclear fuel enrichment, namely plutonium 239 and 241, uranium 235 and uranium 233, the latter has for some time appeared to offer the best economic possibilities for light-water reactors. This is based largely on the fact that the isotope thorium 232 from which uranium 233 is derived by nuclear reaction is in abundant supply. ("Thorium and the Third Fuel"--Joseph M. Dukert--U.S. Atomic Energy Commission.) Not only is thorium more abundant, but uranium 233 when burned as fuel produces more neutrons per atom destroyed than does uranium 235.
Practical aspects and problems of the utilization of uranium 233 for power generation have been explored by one of the instant inventors and set forth in the literature. ("Power Cost Reduction by Cross-Progeny Fueling of Thermal and Fast Reactors"--L. W. Lang, NUCLEAR APPLICATIONS, November 1968; and "Utility Incentives for Implementing Crossed Progeny Fuelings"--L. W. Lang, NUCLEAR APPLICATIONS AND TECHNOLOGY, August 1970.)
While it has been recognized that uranium 233 has many advantages as a fuel for power reactors its use has been accompanied by numerous disadvantages and problems. Most of these stem from the fact that the production of uranium 233 from thorium 232 is usually accompanied by the formation of uranium 232 and its daughter, thorium 228. These isotopes create radiation problems in the separation and fuel preparation processes as well as difficulties in reactor charging. Generally the presence of uranium 232 in over ten parts per million (ppm) in the final fuel embodiment is considered undesirable; under 10 ppm the fuel is termed "clean."
As is well known to those skilled in the art, different reactor concepts produce different results when the thorium cycle is employed. To the present time thorium has been considered best utilized in the high temperature gas cooled reactors (HTGR) and molten-salt reactors (MSR). Its use in light-water reactors (LWR), both pressurized and boiling, has heretofore not been looked upon with much favor. ("The Role of Thorium in Power Reactor Development"--P. R. Kasten, ATOMIC ENERGY REVIEW, IAEA Vienna 1970; "Molten Salt Reactors"--H. G. MacPherson, INTERNATIONAL CONFERENCE, CONSTRUCTIVE USES OF ATOMIC ENERGY, 1966. )
Consideration of the thorium cycle has otherwise been generally confined to breeder reactors. The production of clean uranium 233 for use in power reactors is covered by a patent of one of the instant inventors, namely L. W. Lang, U.S. Pat. No. 3,658,644. This uses a fast breeder reactor (FBR) and novel combinations of reactor core and blanket.
Actually until the time of the present invention a truly clean U-233 enriched reactor fuel having less than ten parts per million of uranium 232 has not been produced commercially for use in practical power reactors. The use of uranium 233 as a reactor fuel has been largely limited to the experimental stage.
All previous systems utilizing thorium as a fertile material in power reactors have certain basic similarities. The thorium is mixed with a fissile driver fuel such as uranium 235. As the uranium 235 fissions the fertile fuel, thorium, is converted to uranium 233 as is well known to those skilled in the art. Continued irradiation, particularly by the fast neutrons from the driver fuel, namely those of an energy content greater than six Mev, create the undesirable isotope, uranium 232, described above. These fast neutron reactions result also from continued irradiation of the uranium 233 which has been formed and fissions in situ. The fissioning of the bred uranium 233 limits the length of irradiation and there remains the hitherto unsolved problem of thorium irradiation in thermal reactors.
Specifically, difficulties in producing "clean" U-233 fuels stem from the fact that thorium 232 can be transformed to protoactinium 231 by extremely energetic neutrons (greater than 6 Mev). The protoactinium 231, by simple neutron capture, forms uranium 232 in the following series of reactions: ##STR1## The U-232 decays with a 74 year half-life to Th-228 which in turn has a 1.9 year half-life and heads a decay chain of short-lived isotopes some of which give off highly energetic beta and gamma radiation. Since chemical separations do not segregate isotopes of the same element, the U-232 is separated with the U-233 and when the former builds up to high levels, handling by remote methods is required. Similarly, the build-up of Th-228 denies re-use of the thorium.
By separating thorium from the fissioning driver fuel the incidence of highly energetic neutrons, required to produce Pa-231 by (n, 2n) Th-232 reaction, is reduced. However, the U-233 formed in situ can in turn fission by thermal neutron absorptions. Thus, the U-233 produced from thermal neutron capture by Th-232 eventually limits the length of the irradiation since its fissioning produces energetic neutrons capable of forming Pa-231.
Chemical processing of reactor fuels is a highly developed art. Thus, it is well known that uranium and thorium can be coprocessed using solvent extraction or similar methods, dividing the uranium and thorium into separate process streams. These separate streams can then be concentrated and converted to solids for fabrication into fuel elements. Where previously irradiated fissionable fuel is involved, provision is made for extracting and disposing of the fission products and for extracting and recycling plutonium as nuclear fuel. Fission element assay methods are available for determining the fissile content of the reactor fuel, both in-core and in subsequent chemical processing.
It is, accordingly, a general object of the invention to provide a thorium fueled light-water cooled nuclear reactor wherein (n,2n) reactions are minimized and, hence, production of U-232 is minimized. It is a further object of the invention to provide for co-processing of the irradiated thorium with uranium so as to produce a U-233 enriched fuel in which the U-232 is diluted to safe handling levels.